B. I. Synzynys, Thi Kim Phung Nguyen, O. Momot, G. Lavrentyeva
The results of the internal radiation dose calculations for the population and the assessment of the radiation risk from radioactive carbon C-14 during the normal operation of the World’s First Nuclear Power Plant in Obninsk and Smolensk NPP are presented. Calculations were carried out using two methods, taking into account the inhalation and oral intake of C-14 with food into the human body. Radiation doses are 5.69·10-9 Sv/year and 5.95·10-9 Sv/year (for Obninsk NPP), 3.77·10-7 Sv/year and 3.96·10-7 Sv/year (for Smolensk NPP), which is orders of magnitude less than the established minimum significant dose (10 μSv). The assessed levels of radiation risk for the population does not exceed the risk established by NRB 99/2009 (1·10-5). It was found that the main contribution to the formation of the dose and risk of internal radiation of the population from C-14 radiation, released by the respective NPP, was the incorporation of radionuclide with locally produced food products, which is confirmed by the results of calculations using two methods, taking into account the influence of two nuclear power plants.
{"title":"Radiation risk assessment for the population from C-14 emissions of the World’s first NPP and Smolensk NPP","authors":"B. I. Synzynys, Thi Kim Phung Nguyen, O. Momot, G. Lavrentyeva","doi":"10.3897/nucet.9.116651","DOIUrl":"https://doi.org/10.3897/nucet.9.116651","url":null,"abstract":"The results of the internal radiation dose calculations for the population and the assessment of the radiation risk from radioactive carbon C-14 during the normal operation of the World’s First Nuclear Power Plant in Obninsk and Smolensk NPP are presented. Calculations were carried out using two methods, taking into account the inhalation and oral intake of C-14 with food into the human body. Radiation doses are 5.69·10-9 Sv/year and 5.95·10-9 Sv/year (for Obninsk NPP), 3.77·10-7 Sv/year and 3.96·10-7 Sv/year (for Smolensk NPP), which is orders of magnitude less than the established minimum significant dose (10 μSv). The assessed levels of radiation risk for the population does not exceed the risk established by NRB 99/2009 (1·10-5). It was found that the main contribution to the formation of the dose and risk of internal radiation of the population from C-14 radiation, released by the respective NPP, was the incorporation of radionuclide with locally produced food products, which is confirmed by the results of calculations using two methods, taking into account the influence of two nuclear power plants.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"50 21","pages":""},"PeriodicalIF":0.0,"publicationDate":"2023-12-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"139007149","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The k eff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.
{"title":"Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code","authors":"Md Imtiaj Hossain, Abdus Sattar Mollah, Yasmin Akter, Mehraz Zaman Fardin","doi":"10.3897/nucet.9.91090","DOIUrl":"https://doi.org/10.3897/nucet.9.91090","url":null,"abstract":"The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for the VVER-1000 MOX core, containing 30% mixed oxide fuel with low enriched uranium fuel, are presented in this study. As per the benchmark report, all six states are considered in the present study. The k eff values, assembly average fission reaction rates, and pin-by-pin fission rates were calculated as per benchmark criteria. In addition, 2D thermal and fast neutron-flux distribution were also generated. The reactivity results and neutron flux distribution were compared with other results in which benchmark analysis was performed using the same core geometry and it showed great similarity with slight deviation. This shows that the modeling of the VVER-1000 MOX core was done successfully using OpenMC. Because OpenMC was successfully used for neutronics calculation of the VVER-1000 whole core, it may be mentioned here that OpenMC code can also be utilized for neutronics and other reactor core physics analyses of the VVER-1200 reactor which is to be commissioned in Bangladesh in the upcoming year.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"117 30","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135136758","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The paper is devoted to refining the Lawson criterion for three scenarios of using D- 3 He fuel in fusion reactors (fully catalyzed and non-catalysed D-D cycles and a D- 3 He cycle with 3 He self-supply). To this end, a new parameterization of the D + 3 He → p + 4 He fusion reaction cross-section and astrophysical factor has been developed based on the effective radius approximation (Landau-Smorodinsky-Bethe approximation), which is a model-free theoretical approach to investigating near-threshold nuclear reactions, including resonant reactions. In the framework of this approximation, experimental data from studies in the NACRE II and EXFOR libraries, believed to provide the most reliable results to date, have been described within the accuracy declared in the studies in question in the energy range of 0 to 1000 keV, and the fusion reactivity averaged over the Maxwell distribution has been calculated. The results obtained are in good agreement with the calculations based on the R -matrix theory and the NACRE II fusion reactivity data. For the fully catalyzed D-D cycle and the cycle with 3 He self-supply, the Lawson criterion and the triple Lawson criterion have been calculated based on solving the equations of the stationary process kinetics in a fusion reactor for three fuel ions (D, 3 He, and T) taking into account the potential for external supply of 3 He and p and 4 He impurity ions removed from the reaction zone. The parameters of the triple Lawson criterion found are as follows: n τ T = 6.42∙10 16 cm -3 ∙s∙keV ( T = 54 keV) for the fully catalyzed D-D cycle, n τ T = 1.03∙10 17 cm -3 ∙s∙keV ( T = 45 keV) for the cycle with 3 He self-supply, and n τ T = 4.89∙10 16 cm -3 ∙s∙keV ( T = 67 keV) for the non-catalyzed D-D cycle with equimolar D- 3 He fuel.
本文致力于改进在聚变反应堆中使用D- 3he燃料的三种情况下的Lawson准则(完全催化和非催化D-D循环以及具有3 - He自供的D- 3he循环)。为此，基于有效半径近似(Landau-Smorodinsky-Bethe近似)，提出了一种新的D + 3 He→p + 4 He聚变反应截面和天体物理因子的参数化方法，这是一种研究近阈值核反应(包括共振反应)的无模型理论方法。在这个近似的框架内，来自naacre II和EXFOR库的实验数据被认为提供了迄今为止最可靠的结果，在0到1000 keV的能量范围内，在研究中所声明的精度范围内进行了描述，并计算了麦克斯韦分布上的平均聚变反应性。所得结果与基于R矩阵理论的计算结果和NACREⅱ型核聚变反应性数据吻合较好。对于完全催化的D-D循环和3he自供循环，在考虑3he的外部供应和p、4he杂质离子从反应区移出的可能性的基础上，通过求解聚变反应堆中三个燃料离子(D、3he和T)的固定过程动力学方程，计算了Lawson准则和三重Lawson准则。发现的三个Lawson判据参数如下:对于完全催化的D-D循环，n τ T = 6.42∙10 16 cm -3∙s∙keV (T = 54 keV)，对于3 He自供的循环，n τ T = 1.03∙10 17 cm -3∙s∙keV (T = 45 keV)，对于等量D- 3 He燃料的非催化D-D循环，n τ T = 4.89∙10 16 cm -3∙s∙keV (T = 67 keV)。
{"title":"Lawson criterion for different scenarios of using D-3He fuel in fusion reactors","authors":"Alexandr I. Godes, Vladimir L. Shablov","doi":"10.3897/nucet.9.114267","DOIUrl":"https://doi.org/10.3897/nucet.9.114267","url":null,"abstract":"The paper is devoted to refining the Lawson criterion for three scenarios of using D- 3 He fuel in fusion reactors (fully catalyzed and non-catalysed D-D cycles and a D- 3 He cycle with 3 He self-supply). To this end, a new parameterization of the D + 3 He → p + 4 He fusion reaction cross-section and astrophysical factor has been developed based on the effective radius approximation (Landau-Smorodinsky-Bethe approximation), which is a model-free theoretical approach to investigating near-threshold nuclear reactions, including resonant reactions. In the framework of this approximation, experimental data from studies in the NACRE II and EXFOR libraries, believed to provide the most reliable results to date, have been described within the accuracy declared in the studies in question in the energy range of 0 to 1000 keV, and the fusion reactivity averaged over the Maxwell distribution has been calculated. The results obtained are in good agreement with the calculations based on the R -matrix theory and the NACRE II fusion reactivity data. For the fully catalyzed D-D cycle and the cycle with 3 He self-supply, the Lawson criterion and the triple Lawson criterion have been calculated based on solving the equations of the stationary process kinetics in a fusion reactor for three fuel ions (D, 3 He, and T) taking into account the potential for external supply of 3 He and p and 4 He impurity ions removed from the reaction zone. The parameters of the triple Lawson criterion found are as follows: n τ T = 6.42∙10 16 cm -3 ∙s∙keV ( T = 54 keV) for the fully catalyzed D-D cycle, n τ T = 1.03∙10 17 cm -3 ∙s∙keV ( T = 45 keV) for the cycle with 3 He self-supply, and n τ T = 4.89∙10 16 cm -3 ∙s∙keV ( T = 67 keV) for the non-catalyzed D-D cycle with equimolar D- 3 He fuel.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"114 10","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-11-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135137257","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Evgenia A. Kuleshova, Ivan V. Fedotov, Dmitry A. Maltsev, Margarita G. Isaenkova, Olga A. Krymskaya, Roman A. Minushkin
The paper considers the results of structural studies and mechanical tests after a long-term thermal exposure of laboratory heats of the metallurgically improved 15Kh2NMFA steel and steel with an increased content of nickel considered as materials for the pressure vessels of advanced VVER-type reactors of various designs. It has been shown that, both for the improved 15Kh2NMFA steel and the high-nickel steel, there are no signs of grain boundary embrittlement after an segregation provoking embrittlement heat treatment. This is explained by the extremely low grain boundary segregation of phosphorus in the initial state caused by a high degree of the structure dispersity as well as by rather a low content of impurities. Besides, no changes have been found in the yield strength value for the improved 15Kh2NMFA steel, which agrees with the structure investigation results. For the high-nickel steel, a tendency towards a minor yield strength decrease by 5 to 10% and a regular reduction of the critical brittleness temperature has been revealed. A decrease in the mechanical properties has been caused by a relatively low temperature of tempering for the high-nickel steel and, accordingly, by the potential occurrence of the structure recovery during long-term thermal exposure, as evidenced by the results of an X-ray diffraction analysis. Despite the structure recovery in the high-nickel steel under the long-term thermal exposure, the main strengthening carbide phases remain stable. Due to this, the yield strength value remains at a relatively high level that exceeds the values for the modern VVER-type vessel steels, even in the case of a thermal exposure much in excess of the expected operating conditions for advanced VVER reactors. The observed decrease of critical brittleness temperature during the long-term thermal exposure contributes to an increase in the steel resistance to brittle fracture.
{"title":"Thermal resistance of steels with increased strength properties for pressure vessels of advanced VVER reactors of various designs","authors":"Evgenia A. Kuleshova, Ivan V. Fedotov, Dmitry A. Maltsev, Margarita G. Isaenkova, Olga A. Krymskaya, Roman A. Minushkin","doi":"10.3897/nucet.9.113715","DOIUrl":"https://doi.org/10.3897/nucet.9.113715","url":null,"abstract":"The paper considers the results of structural studies and mechanical tests after a long-term thermal exposure of laboratory heats of the metallurgically improved 15Kh2NMFA steel and steel with an increased content of nickel considered as materials for the pressure vessels of advanced VVER-type reactors of various designs. It has been shown that, both for the improved 15Kh2NMFA steel and the high-nickel steel, there are no signs of grain boundary embrittlement after an segregation provoking embrittlement heat treatment. This is explained by the extremely low grain boundary segregation of phosphorus in the initial state caused by a high degree of the structure dispersity as well as by rather a low content of impurities. Besides, no changes have been found in the yield strength value for the improved 15Kh2NMFA steel, which agrees with the structure investigation results. For the high-nickel steel, a tendency towards a minor yield strength decrease by 5 to 10% and a regular reduction of the critical brittleness temperature has been revealed. A decrease in the mechanical properties has been caused by a relatively low temperature of tempering for the high-nickel steel and, accordingly, by the potential occurrence of the structure recovery during long-term thermal exposure, as evidenced by the results of an X-ray diffraction analysis. Despite the structure recovery in the high-nickel steel under the long-term thermal exposure, the main strengthening carbide phases remain stable. Due to this, the yield strength value remains at a relatively high level that exceeds the values for the modern VVER-type vessel steels, even in the case of a thermal exposure much in excess of the expected operating conditions for advanced VVER reactors. The observed decrease of critical brittleness temperature during the long-term thermal exposure contributes to an increase in the steel resistance to brittle fracture.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135569916","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Daniil M. Arkhangelsky, Yuliya S. Daichenkova, Mikhail A. Kalugin, Dmitry S. Oleynik, Denis A. Shkarovsky
The paper considers a computational study of the importance function effect on the accuracy of calculating the effective fraction of delayed neutrons, β eff , and generation time of instantaneous neutrons using the MCU Monte Carlo code based on the example of three criticality experiments from the ICSBEP handbook. In the MCU code, the importance function has a piecewise constant form: the computational model is broken down into a finite number of registration objects, and the neutron importance is calculated in each. The obtained importance values are used then to calculate the kinetic functionals due to which the calculation accuracy for the latter depends on the resolution. Three types of the importance function spatial partition (axial, radial, combined) have been studied. The numerical simulation results have shown that the axial component of the neutron importance function in all experiments has practically no effect on the calculation accuracy for β eff and Λ: the difference between the obtained values is less than 1%. The radial component has a notable effect (of up to 15.9%) on the Λ calculation accuracy while having almost no effect on the β eff estimate. Using combined partition, as compared with radial partition, improves the calculation accuracy insignificantly (< 1%).
{"title":"The effect of the importance function resolution on the accuracy of calculating the functionals of the neutron kinetics in water critical assemblies by Monte Carlo method","authors":"Daniil M. Arkhangelsky, Yuliya S. Daichenkova, Mikhail A. Kalugin, Dmitry S. Oleynik, Denis A. Shkarovsky","doi":"10.3897/nucet.9.112165","DOIUrl":"https://doi.org/10.3897/nucet.9.112165","url":null,"abstract":"The paper considers a computational study of the importance function effect on the accuracy of calculating the effective fraction of delayed neutrons, β eff , and generation time of instantaneous neutrons using the MCU Monte Carlo code based on the example of three criticality experiments from the ICSBEP handbook. In the MCU code, the importance function has a piecewise constant form: the computational model is broken down into a finite number of registration objects, and the neutron importance is calculated in each. The obtained importance values are used then to calculate the kinetic functionals due to which the calculation accuracy for the latter depends on the resolution. Three types of the importance function spatial partition (axial, radial, combined) have been studied. The numerical simulation results have shown that the axial component of the neutron importance function in all experiments has practically no effect on the calculation accuracy for β eff and Λ: the difference between the obtained values is less than 1%. The radial component has a notable effect (of up to 15.9%) on the Λ calculation accuracy while having almost no effect on the β eff estimate. Using combined partition, as compared with radial partition, improves the calculation accuracy insignificantly (&lt; 1%).","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"21 5 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135569906","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.
{"title":"Evaluation of neutronic performance for the VVER-1000 reactor core with regenerated uranium-plutonium fuel","authors":"Viktor V. Semishin","doi":"10.3897/nucet.9.112325","DOIUrl":"https://doi.org/10.3897/nucet.9.112325","url":null,"abstract":"The possibility has been considered for the VVER-1000 reactor fuel loading to be formed based on regenerated fuel with the use of the spent fuel accumulated in reactors of the same type. A study was undertaken to investigate the change in the isotopic composition of the plutonium discharged from a thermal reactor in the course of its multiple reprocessing and recycle in a thermal neutron reactor. To obtain an equilibrium isotopic composition of the reactor-grade plutonium, 3D neutronic calculations were performed for the stationary fuel cycles of a VVER-1000 serial reactor with conventional oxide fuel and oxide fuel based on regenerated uranium and based on an undivided mixture of uranium and plutonium oxides from SNF. The neutronic performance of reactor cores was compared for the above mentioned fuel types in the course of the fuel company, including the following: in-core radial power density shaping, values of reactivity coefficients for various thermal parameters, reactivity control system efficiency, etc.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"67 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135617100","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The publication substantiates the economic advantages of using in the starting loads of inherently safe fast reactors with a closed fuel cycle of enriched uranium instead of uranium-plutonium regenerate obtained by reprocessing of thermal reactors spent nuclear fuel (SNF). The justifications are given taking into account both the preliminary technical and economic assessments carried out by the basic enterprises of TVEL JSC and SHK JSC, and the neutron-physical and system-economic studies performed at the Private Institution of the ITCP Proryv (Breakthrough). It is shown that the starting-up of a fast reactor on enriched uranium instead of uranium-plutonium fuel, taking into account the costs of preliminary reprocessing of thermal reactors spent fuel, allows achieving a significant economic gain at the stage of construction and commissioning of nuclear power plants. It is also shown that even at moderately high values of the discount coefficient, the uranium start of a fast reactor with a closed fuel cycle is economically preferable in comparison with the option of starting on uranium-plutonium fuel from the positions of the break-even tariff.
{"title":"Economic advantages of starting up of inherently safe fast reactors with a closed fuel cycle on fortificated uranium","authors":"Mikhail A. Orlov","doi":"10.3897/nucet.9.111914","DOIUrl":"https://doi.org/10.3897/nucet.9.111914","url":null,"abstract":"The publication substantiates the economic advantages of using in the starting loads of inherently safe fast reactors with a closed fuel cycle of enriched uranium instead of uranium-plutonium regenerate obtained by reprocessing of thermal reactors spent nuclear fuel (SNF). The justifications are given taking into account both the preliminary technical and economic assessments carried out by the basic enterprises of TVEL JSC and SHK JSC, and the neutron-physical and system-economic studies performed at the Private Institution of the ITCP Proryv (Breakthrough). It is shown that the starting-up of a fast reactor on enriched uranium instead of uranium-plutonium fuel, taking into account the costs of preliminary reprocessing of thermal reactors spent fuel, allows achieving a significant economic gain at the stage of construction and commissioning of nuclear power plants. It is also shown that even at moderately high values of the discount coefficient, the uranium start of a fast reactor with a closed fuel cycle is economically preferable in comparison with the option of starting on uranium-plutonium fuel from the positions of the break-even tariff.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135617711","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Artem Z. Gayazov, Oleg Z. Gaiazov, Viacheslav Yu. Kozlov, Sergey V. Pavlov, Aleksandr A. Samsonov
One of the safety requirements regarding the shipping cask for spent nuclear fuel is that its leak-tightness should be maintained by preserving the cask body structural integrity and the sealing system tightness under normal and accident transportation conditions. The cask under design has a cylindrical process penetration (port) in its bottomб which is sealed using a plug with a radial seal composed of two rubber O-rings. The cask sealing assembly design was justified by the ANSYS LS-DYNA code calculation results. In particular, the strains of the cask components were calculated when dropped from a height of 1 m with the sealing assembly hitting a vertical bar. The cask was concluded to be leak-tight or leaky based on the strain nature and amount. To verify the adequacy of the results, computer-aided and realistic simulations were undertaken with a 1/2.5 scale mockup cask dropped on a bar from a height of 1 m. The computational and experimental results show a good agreement in terms of the impact response accelerations (overloads) for the mockup cask and bar collision and in terms of the plastic strains for the key components of the mockup bottom port sealing assembly. This proves the adequacy of the numerical cask model that has been developed and the efficiency of the LS-DYNA simulations. The inner rubber O ring compression is reduced by the plastic strains in the cask’s bottom port area, leading to a loose inner radial seal, as shown by the calculations. But the outer seal remains leak-tight, ensuring so the mockup cask tightness. The physical test results have also confirmed that the mockup cask remains leak-tight.
{"title":"Calculation and experimental studies for the spent nuclear fuel shipping cask sealing assembly","authors":"Artem Z. Gayazov, Oleg Z. Gaiazov, Viacheslav Yu. Kozlov, Sergey V. Pavlov, Aleksandr A. Samsonov","doi":"10.3897/nucet.9.113520","DOIUrl":"https://doi.org/10.3897/nucet.9.113520","url":null,"abstract":"One of the safety requirements regarding the shipping cask for spent nuclear fuel is that its leak-tightness should be maintained by preserving the cask body structural integrity and the sealing system tightness under normal and accident transportation conditions. The cask under design has a cylindrical process penetration (port) in its bottomб which is sealed using a plug with a radial seal composed of two rubber O-rings. The cask sealing assembly design was justified by the ANSYS LS-DYNA code calculation results. In particular, the strains of the cask components were calculated when dropped from a height of 1 m with the sealing assembly hitting a vertical bar. The cask was concluded to be leak-tight or leaky based on the strain nature and amount. To verify the adequacy of the results, computer-aided and realistic simulations were undertaken with a 1/2.5 scale mockup cask dropped on a bar from a height of 1 m. The computational and experimental results show a good agreement in terms of the impact response accelerations (overloads) for the mockup cask and bar collision and in terms of the plastic strains for the key components of the mockup bottom port sealing assembly. This proves the adequacy of the numerical cask model that has been developed and the efficiency of the LS-DYNA simulations. The inner rubber O ring compression is reduced by the plastic strains in the cask’s bottom port area, leading to a loose inner radial seal, as shown by the calculations. But the outer seal remains leak-tight, ensuring so the mockup cask tightness. The physical test results have also confirmed that the mockup cask remains leak-tight.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"73 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135569907","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Vladimir A. Gorbunov, Svetlana S. Teplyakova, Nikita A. Lonshakov, Sergey G. Andrianov, Pavel A. Mineev
The paper presents the results of a numerical study into the efficiency of the fuel element operation in the pressurized water reactor (VVER) core filled with uranium dioxide (UO 2 ) pellets. The investigation results were obtained from a three-dimensional simulation of the fuel element power density. The dependencies of the fuel and fuel cladding temperatures on specific power per cubic meter of fuel are compared. Uranium metal and uranium dioxide have been studied as fuel. Engineering constraints on the safe operation of fuel assemblies have been selected as the determining parameters. The paper analyzes the extent of the radiation heat transfer effects on the fuel element specific power. Equations have been obtained that reflect the dependencies of specific power per cubic meter of fuel on the size of the fuel pellet hole diameter in the maximum heat flux conditions. The COMSOL Multiphysics code, a numerical thermophysical simulation package, was used for the study. Calculations show that an additional uranium-235 enrichment with an increase in the fuel pellet hole diameter at a fixed fuel thermal power leads to a reduced reactor axial temperature field peaking factor.
{"title":"Investigation of the influence of the fuel element design parameter on the VVER-1000 reactor axial power peaking factor","authors":"Vladimir A. Gorbunov, Svetlana S. Teplyakova, Nikita A. Lonshakov, Sergey G. Andrianov, Pavel A. Mineev","doi":"10.3897/nucet.9.113622","DOIUrl":"https://doi.org/10.3897/nucet.9.113622","url":null,"abstract":"The paper presents the results of a numerical study into the efficiency of the fuel element operation in the pressurized water reactor (VVER) core filled with uranium dioxide (UO 2 ) pellets. The investigation results were obtained from a three-dimensional simulation of the fuel element power density. The dependencies of the fuel and fuel cladding temperatures on specific power per cubic meter of fuel are compared. Uranium metal and uranium dioxide have been studied as fuel. Engineering constraints on the safe operation of fuel assemblies have been selected as the determining parameters. The paper analyzes the extent of the radiation heat transfer effects on the fuel element specific power. Equations have been obtained that reflect the dependencies of specific power per cubic meter of fuel on the size of the fuel pellet hole diameter in the maximum heat flux conditions. The COMSOL Multiphysics code, a numerical thermophysical simulation package, was used for the study. Calculations show that an additional uranium-235 enrichment with an increase in the fuel pellet hole diameter at a fixed fuel thermal power leads to a reduced reactor axial temperature field peaking factor.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"44 25 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135616967","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Gleb W. Karpovich, Yury A. Kazansky, Kirill A. Bakhantsov, Kirill A. Isanov, Nikita O. Kushnir
The paper deals with assessing the effects of the ENDF/B-VI.8, ENDF/B-VII.0, JEFF 3.1 and JEFF 3.1.1 nuclear data libraries on the results of calculating a number of functionals for a system based on a VVER reactor with fuel with a large fraction of minor actinides (up to 10%). Key estimates have been obtained for the errors introduced by libraries in calculations of systems with minor actinides (MA) based on a VVER-1200 reactor: – for reactivity, σ ρ = 0.3 β eff ; – for isotopic compositions with minor actinides, ≤ 5% (the error for each particular isotope is different); – for the total mass of accumulated MAs, ε m = 0.8%. Conclusions have been made with respect to the need for the further refinement of the library MA data proceeding from the nature of the calculation tasks that dictate the requirements for the accuracy of nuclear constants. It has been shown that systems based on VVER-1000/1200/1300 reactors with MAs need to be calculated using several libraries of evaluated nuclear data created at different organizations and based on the largest possible number of non-recurrent sets of experimental data.
{"title":"Effects of evaluated nuclear data libraries on the calculation results for fuel burnup with minor actinides in a VVER reactor","authors":"Gleb W. Karpovich, Yury A. Kazansky, Kirill A. Bakhantsov, Kirill A. Isanov, Nikita O. Kushnir","doi":"10.3897/nucet.9.112327","DOIUrl":"https://doi.org/10.3897/nucet.9.112327","url":null,"abstract":"The paper deals with assessing the effects of the ENDF/B-VI.8, ENDF/B-VII.0, JEFF 3.1 and JEFF 3.1.1 nuclear data libraries on the results of calculating a number of functionals for a system based on a VVER reactor with fuel with a large fraction of minor actinides (up to 10%). Key estimates have been obtained for the errors introduced by libraries in calculations of systems with minor actinides (MA) based on a VVER-1200 reactor: – for reactivity, σ ρ = 0.3 β eff ; – for isotopic compositions with minor actinides, ≤ 5% (the error for each particular isotope is different); – for the total mass of accumulated MAs, ε m = 0.8%. Conclusions have been made with respect to the need for the further refinement of the library MA data proceeding from the nature of the calculation tasks that dictate the requirements for the accuracy of nuclear constants. It has been shown that systems based on VVER-1000/1200/1300 reactors with MAs need to be calculated using several libraries of evaluated nuclear data created at different organizations and based on the largest possible number of non-recurrent sets of experimental data.","PeriodicalId":100969,"journal":{"name":"Nuclear Energy and Technology","volume":"23 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135617109","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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